Assessment of Gamma Ray Shielding Properties for Skin

Malidarre, Roya Boodaghi; Tekın, H.O.; Günoğlu, Kadir; Akyıldırım, Hakan

doi:10.22399/ijcesen.1247867

Cited by 25 publications

(2 citation statements)

References 22 publications

(23 reference statements)

Supporting

Mentioning

Contrasting

Order By: Relevance

“…In the literature, MCNP has been used to calculate the radiation protection performance of various materials, such as concretes [42], glasses [43], and metal composites [33]. Simulation calculations for radiation shielding calculations are widely used in the literature [44][45][46][47].…”

Section: Monte Carlo Simulationsmentioning

confidence: 99%

Investigating Thermal Neutron and Gamma Ray Shielding Properties of Al Matrix Gd2O3-and W-Doped Composites Using Monte Carlo Simulations

Gaylan,

Bozkurt

2024

Süleyman Demirel Üniversitesi Fen Edebiyat Fakültesi Fen Dergisi

View full text Add to dashboard Cite

This study aimed to calculate the thermal neutron (0.0253 eV) total macroscopic cross-section and gamma-ray (1.25 MeV) linear absorption coefficient for (100-x)Al-xGd2O3 (x=5 to 50) and 100-(x+y)Al-xGd2O3+yW (x,y=5 to 50) composites using MCNP6.2 simulation code. The simulation consists of a mono-energy point neutron and gamma-ray source, target material, and detector. The F4 tally from the MCNP6.2 library was used as the detector. The results show the highest thermal neutron total macroscopic cross-section for the 50%Al-50%Gd2O3 composite and the highest linear absorption coefficient for the 50%Al-%5Gd2O3-45%W composite. The results of this study provide a good understanding of the thermal neutron and gamma ray the shielding capabilities of Al matrix Gd2O3 and W doped composites.

show abstract

Section: Monte Carlo Simulationsmentioning

confidence: 99%

Investigating Thermal Neutron and Gamma Ray Shielding Properties of Al Matrix Gd2O3-and W-Doped Composites Using Monte Carlo Simulations

Gaylan,

Bozkurt

2024

Süleyman Demirel Üniversitesi Fen Edebiyat Fakültesi Fen Dergisi

View full text Add to dashboard Cite

show abstract

“…However, it is important to remember that uncontrolled exposure to strong ionizing radiation sources can easily penetrate living tissue and cause severe damage to the human body, resulting in tumors or death. Therefore, effective radiation protection techniques are required to ensure that the benefits of nuclear energy are reaped without harming staff (9)(10)(11)(12)(13).…”

Section: Introductionmentioning

confidence: 99%

Advances in Polymeric Neutron Shielding: The Role of Benzoxazine-h-BN Nanocomposites in Nuclear Protection

et al. 2023

View full text Add to dashboard Cite

Given their substantial neutron capture cross-section, extreme hardness, and high chemical and thermal stability, boron-based materials are widely used as building blocks to protect against highly ionizing radiations such as gamma rays and neutrons. Indeed, uncontrolled nuclear radiation exposure can be highly hazardous to radiation workers and the public. In this sense, this work presents an extensive study and experimental evaluation of the nuclear shielding features of hexagonal-boron nitride (h-BN) based nanocomposite, where bisphenol-A based polybenzoxazine (BA-PBz) was used as matrix. The neutron shielding studies were carried out at the nuclear research reactor of Algeria NUR. The surface treatment of h-BN nanoparticles was confirmed by FTIR and XPS techniques. The curing behavior and the degradation phenomena of the nanocomposites were evaluated by DSC-TGA analyses. The distribution of h-BN nanoparticles within the polymer matrix was assessed by TEM and SEM. The results showed that the developed boron nitride-based nanocomposite exhibits intriguing shielding performances and good thermal stability. The DSC-TGA tests exhibit high degradation temperature that reach 279°C. The highest performances were obtained at an h-BN concentration of 7 wt%, where the macroscopic cross was found to be (Σ = 3.844 cm–1) with a screening ratio of (S = 96.12%), equivalent to a mean free path (λ) of 0.138 cm.

show abstract